A nuclear light water reactor of boiling water type includes a plurality of elongated fuel units, which contain fissible material, and a number of control rods. The fuel units may be designed as elongated fuel assemblies including a number of fuel rods each having a tubular cladding enclosing a pile of fissible fuel. In a boiling water reactor there is large number of such fuel assemblies, in the order of 400 to 800, and approximately a fourth of control rods, i.e. in the order of 100 to 200.
The fuel units are arranged in parallel to each other and grouped in a plurality of cells which each may include four fuel units. These cells form together the core of the reactor. Substantially each such cell in the core includes a control rod position. In each of these control rod positions, one of the control rods is completely or partly introduceble. The control rods contain neutron absorbing material, such as boron or hafnium, and are used in a boiling water reactor for controlling and interrupting the nuclear reaction in the fuel. When all control rods are introduced, the reactor is shut down, i.e. more neutrons than being released in the fission process will be absorbed and the nuclear reaction decays.
The fuel units, comprised in the core during an operation cycle, are different with regard to the amount of fissible material. This difference depends in the first place on the fact that the fuel units have been in operation during different time periods. A first type of fuel units may be the new ones thus including a relatively large amount of fissible material. A second type of fuel units may have a certain degree of burn out obtained during one or more preceding operation cycles in a reactor. This second type of fuel units thus includes a relatively smaller amount of fissible material. The fuel units may also from the beginning be designed with different amount and distribution of the fissible material.
During an operation cycle the different types of fuel units are arranged in such a way that they are distributed and mixed in the core. The fuel units containing new fuel are preferably located in the proximity of the centre of the core whereas the fuel units having the largest burn out degree, i.e. the smallest amount of fissible material, preferably are located in the proximity of the periphery of the core. This reduces the leakage of neutrons out from the core and is economically advantageous, but also result in a higher effect and greater thermal loads on the centrally located fuel.
The control rods may be divided into various groups, for instance shut down rods, which merely are introduced in the core when the reactor is shut down, and controlling control rods used for controlling the effect of the reactor. Before the reactor is started and an operation cycle is initiated, substantially all control rods are introduced in the core. When the operation cycle begins, a majority of the control rods, for instance about 90%, are extracted from the core. During normal operation of the reactor approximately a tenth of the control rods are thus completely or partly introduced in the core. The primary purpose with the controlling control rods, which are introduced during operation of the reactor, is to absorb excess reactivity in the core. The excess reactivity is built into the core to be successively consumed during the operation cycle, the length of which may vary significantly from less than 12 months to more than 24 months. A long operation cycle also requires correspondingly greater excess reactivity. Such an excess reactivity is accomplished by a larger part of the fuel being new and thus containing a higher concentration of fissible material.
A secondary function of the controlling control rods is to control the effect distribution in the core, partly in such a way that no thermal limits are locally exceeded and partly in such a way that the burn out of fissible material is distributed so that no locally high effects arises when the control rods at the end of the operation cycle have to be extracted when the excess reactivity decreases. It is then required that merely the distribution of the fissible material can control the effect distribution. In this controlling function the control rods co-act with the initial distribution of fissible material and burnable absorbers, see below, which is co-optimised with calculations before each new operation cycle.
The control rods are not themselves sufficient for absorbing all excess reactivity, especially not during operation cycles longer than 12 months. As a supplement burnable absorbers, for instance Gd2O3, which is fixedly included in the new fuel, are therefore provided. Such a burnable absorber is dimensioned to be burnt-out during the first operation cycle. The burnable absorbers also supplement the control of the effect distribution of the core.
The control rods may also be divided into different groups depending on with which cells they are intended to co-act. The control rods may then include first control rods, which co-act with cells with one or several of the first type of fuel units with relatively new fuel, and second control rods, which co-act with cells with the second type of fuel units with partly burn out fuel. The uneven concentration of fissible material in the core, which depends on the fact that the core includes fuel with different degree of burn out, creates problems when determining which control rods are to be introduced during various phases of the operation cycle. The fuel units, which are located most closely to an introduced control rod, will not be burnt-out to the same extent as the fuel units which are located at a distance from this control rod. The relatively small number of control rods in the core during operation thus leads successively to an increasing uneveness in the concentration of fissible material in the core. In addition, a relatively large effect increase is obtained in the fuel units located most closely to the actual control rod position immediately after the control rod has been extracted from the core. Such an effect increase can lead to so called PCI-defects (Pellet Cladding Interaction).
PCI, i.e. mechanical interaction between the pellet and the cladding, which via stress corrosion from fission products leads to a crack on the cladding from inside, is a now well investigated defect mechanism which is described in the specialist literature. For a defect to arise several conditions have to be obtained simultaneously:    1. The burn out is to be sufficiently high so that there is a sufficient amount of fission products, so that the cladding is irradiation hardened and so that there is mechanical contact between the pellet and the cladding. With the actual rod design this occurs at a burn out or 15-20 MWd/kgU. Approximately this is valid for about 60% of the core in the beginning of the operation cycle and for about 80% of the core at the end of the operation cycle.    2. The effect increase has to be so quick that the cladding material does not have time to creep and to reduce the stress level. At the first start after a reloading this is valid for a large part of the core, but the reloading is normally performed with the conditioning rules that have shown to be very efficient. During an operation cycle there are then only preconditions for sufficiently large and quick effect increases beside a controlling control rod that are manipulated during the operation cycle.    3. The end effect has to be sufficiently high, partly for the same reasons as for point 1.    4. The high stress level has to be maintained during a sufficiently long time period in order to permit the stress corrosion to act. From tests the required time period is judged to be from a 10 minutes to several hours. Sufficient durability (holding time) is always present in connection with stationary operation, however not at transients.    5. To these conditions it should also be added that local defect notches from, for instance pellet fragments from the manufacturing or cracking during operation, appear to be necessary. Both operation experiences and ramp tests show a significant distribution which hardly may be explained in any other way.
These conditions are well proved empirically and PCI is generally regarded as an eliminated defect cause through more careful operation rules with slow effect increase (conditioning), through decreased longitudinal heat load (more and thinner fuel rods) and through different variants of Zr-liners (inner layer of soft, low-alloyed Zr on the inner side of the cladding tube). No protection is however 100% safe and it is important to introduce new operation modes in such a way that the risks are not unnecessarily increased. In this context it is also important to note that the PCI-stresses are significantly higher at the extraction of the control rods than at the introduction. The difference may be a factor 10.
The fuel units beside the control rods does not only obtain a lower average burn out but also a skewed burn out since the fuel rods most closely to the control rods are burnt-out very slowly whereas fissible plutonium is generated at a substantially normal degree in these fuel rods. When the control rod is extracted after a long time period of operation, a skewed distribution of fissible material has thus been formed with a corresponding skewed effect distribution as a result, which means that the thermal margin is deteriorated.
These problems may according to the prior art be solved in various ways. According to one known method one may during an operation cycle change the control rod configuration at relatively short intervals according to a predetermined sequence. Such a method is suggested in U.S. Pat. No. 3,385,758.
A disadvantage of this known solution is that after a certain time of the operation cycle it may be difficult to find new control rod configurations with suitable positions for the control rods. Several disadvantages with this known method are described in U.S. Pat. No. 4,285,769, for instance that the reactor effect has to be decreased at each change of the control rod configuration. The factor of capacity, i.e. the average effect production capability of the reactor is decreased.
U.S. Pat. No. 4,285,769 instead suggests that the core is divided into two different types of cells. The first type contains fuel assemblies with relatively new fuel with high reactivity and the second type contains fuel assemblies with partly burnt-out fuel with low reactivity. According to the method defined in U.S. Pat. No. 4,285,769 no control rods are introduced in the cells of the first type but all control takes place in that the control rods are introduced into a part of the cells of the second type. In such a way at least a part of the previously necessary control rod movements may be avoided.
These known methods for controlling the control rods during operation are insufficient when the operation cycles become longer. They have been excellent at the relatively short operation cycles which previously have been used, i.e. an operation cycle of up to 1 year or in a best case in certain applications up to 1.5 years. It is now more common with longer operation cycles, i.e. up to 2 years. At such operation cycles with correspondingly higher excess reactivity new strategies are required for controlling the control rods.
Further examples of control rod strategies are described in the following documents.
U.S. Pat. No. 4,368,171 describes a method for controlling a nuclear reactor by means of control rods in order to obtain a more uniform radial effect distribution. The control rods are divided into different groups at different radial distance from the centre of the core.
U.S. Pat. No. 5,217,678 describes another method for controlling a nuclear reactor by means of control rods which are positionable in different control rod patterns. This known method concerns the control of the control rods during change from one control rod pattern to another control rod pattern.
U.S. Pat. No. 5,307,387 describes a method for loading fuel assemblies in a core in a reactor. The method is characterised in that peripherally located fuel assemblies are positioned in a central part of the core after at least two operation cycles.
U.S. Pat. No. 5,677,938 describes a further method for operating a nuclear reactor. The core is divided into a central area, an intermediate area and a peripheral area. The control rods are grouped in different groups which each is distributed over the whole core. The different control rod groups are introduced after each other at least partly in the core during a desired time interval. This time interval is equally long for all control rod groups.